ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 40

Document Title:

Lawrence Livermore National Laboratory 

– Occupational Internal Dose

Document Number: ORAUT-TKBS-0035-5 

Revision: 00

Effective Date: 10/07/2005

Type of Document: TBD

Supersedes: None

Subject Experts: Carol D. Berger and Paul Szalinski

Document Owner

Approval: Signature on File Approval Date: 09/29/2005

Jay J. Maisler, TBD Team Leader

Approval: Signature on File Approval Date: 09/29/2005

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 09/29/2005

Richard E. Toohey, Project Director

Approval: Signature on File Approval Date: 10/07/2005

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 2 of 40

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

10/07/2005 00 New technical basis document for the Lawrence Livermore National 

Laboratory – Occupational Internal Dose. First approved issue. 

Training required: As determined by the Task Manager. Initiated by 

Jay J. Maisler.
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 3 of 40

TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations ...............................................................................................................

PAGE

5

5.1 Introduction................................................................................................................................6

5.2 Source Term ..............................................................................................................................8

5.3 Measurement Methods ..............................................................................................................8

 5.3.1 Measurement Types and Detection Levels ................................................................. 15

 5.3.2 Reporting Formats and Codes....................................................................................18

 5.3.3 Assessment of Intake for Unmonitored Claimants ......................................................19

5.4 Significant Incidents with Internal Dose Potential ...................................................................... 19

References ........................................................................................................................................21

Glossary ............................................................................................................................................24

ATTACHMENT A, SOURCE TERM BY BUILDING ...........................................................................27
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 4 of 40

LIST OF TABLES

TABLE TITLE PAGE

5-1 Source term by building ...........................................................................................................9

5-2 Isotopic information for aging of weapons-grade plutonium.................................................... 14

5-3 Isotopic information for various uranium enrichment levels .................................................... 15

5-4 In vivo and in vitro sampling frequencies................................................................................ 16

5-5 Bioassay detection levels....................................................................................................... 17

5-6 Bioassay record codes........................................................................................................... 18

5-7 Input parameters for significant incidents ............................................................................... 20

A-1 Source term by building ......................................................................................................... 27
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 5 of 40

ACRONYMS AND ABBREVIATIONS

AMAD activity median aerodynamic diameter

AVLIS Atomic Vapor Laser Isotope Separation

CEDE committed effective dose equivalent

CFR Code of Federal Regulations

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

F fast (solubility rate)

HEPA high-efficiency particulate air (filter)

HEU highly enriched uranium

HPGe hyperpure germanium

ICP-MS inductively coupled plasma-mass spectroscopy

ICRP International Commission on Radiological Protection

IRF intake retention fraction

KPA kinetic phosphorescence analysis

LINAC linear accelerator

LLNL Lawrence Livermore National Laboratory

LSC liquid scintillation counting

M moderate (solubility rate)

MDA minimum detectable amount

MDC minimum detectable concentration

M&E Mechanical and Electrical (Engineering Divisions)

MFP mixed fission products

NESHAP National Emission Standard for Hazardous Air Pollutants

NIOSH National Institute for Occupational Safety and Health 

R&D Research and Development

S slow (solubility rate)

TEDE total effective dose equivalent

TRU transuranic

U.S.C. United States Code

WB whole-body
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 6 of 40

5.1 INTRODUCTION

Site Profiles, which include Technical basis documents (TBDs), are not official determinations made 

by the National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised when additional 

relevant information is obtained about the affected site(s). These documents may be used to assist 

the NIOSH staff in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building or group of buildings 

that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act of 2000 (42 U.S.C. § 7384l (5) and (12)). EEOICPA 

defines a DOE facility as “any building, structure, or premise, including the grounds upon which such 

building, structure, or premise is located ... in which operations are, or have been, conducted by, or 

on behalf of, the Department of Energy (except for buildings, structures, premises, grounds, or 

operations ... pertaining to the Naval Nuclear Propulsion Program).” 42 U.S.C. § 7384l(12). 

Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted above, any 

facility that performs or performed DOE operations of any nature whatsoever is a DOE facility 

encompassed by EEOICPA.

For DOE employees with cancer, the DOE facility definition only determines eligibility for a dose 

reconstruction, which is a prerequisite to a compensation decision (except for members of the Special 

Exposure Cohort). The compensation decision for cancer claimants is based on a section of the 

statute entitled “Exposure in the Performance of Duty.” That provision (42 U.S.C. § 7384n(b)) says 

that an individual with cancer “shall be determined to have sustained that cancer in the performance 

of duty for purposes of the compensation program if, and only if, the cancer ... was at least as likely 

as not related to employment at the facility [where the employee worked], as determined in 

accordance with the [probability of causation] guidelines established under subsection (c) ...” 42 

U.S.C. § 7384n(b). Neither the statute nor the probability of causation guidelines (nor the dose 

reconstruction regulation) define “performance of duty” for DOE employees with a covered cancer or 

restrict the “duty” to nuclear weapons work.

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 

U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program.” 42 U.S.C. § 7384l(12). 

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

(i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”) does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all radiation exposures in its dose 

reconstructions for employees at DOE facilities, including radiation exposures related to the Naval 

Nuclear Propulsion Program. As a result, all internal and external dosimetry results are considered 

valid for use in dose reconstruction. No efforts are made to determine the eligibility of any fraction of 

total measured exposure for inclusion in dose reconstruction.

This Site Profile documents historical practices at the Lawrence Livermore National Laboratory (LLNL) 

and provides information for the evaluation of internal and external dosimetry data for unmonitored 

and monitored workers; it can serve as a supplement to, or substitute for, individual monitoring data.

This document provides a uniform and consistent approach to assessing occupational internal dose at 

the Lawrence Livermore National Laboratory (LLNL) for dose reconstructions for NIOSH in relation to 

the EEOICPA. It provides guidance to dose reconstructors on input parameters that are specific to 
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 7 of 40

LLNL employees, as well as the approach for employees with either missing or no monitoring 

information.

LLNL workers handled a variety of radionuclides as part of their routine work. The key elements in 

the source term were plutonium and tritium, although others were used at various times and in various 

forms. For the purposes of dose reconstruction it can be assumed that internal source terms were 

introduced at LLNL’s inception on September 2, 1952.

Prior to the early 1960s, the only methodology that LLNL used to monitor employees for intakes of 

radionuclides at LLNL was urine bioassay, with the primary focus on excreted tritium. It is not clear 

when bioassay monitoring first began, but it continues to this day for plutonium, americium, uranium, 

mixed fission products, a variety of tracer radionuclides, iodine, and tritium (Mansfield 2000). Air 

monitoring in workplaces and in the breathing zones of employees has been a common surveillance 

method. However, LLNL apparently did not use the data acquired through that program, which are 

not readily associated with individual exposures, to prepare the dose of record for employees.

In vivo methodologies began on an investigational basis in 1964, focused on high-energy (i.e., 

greater-than-200-keV) gamma emitters. In the 1970s, there were attempts to detect low-energy 

photon emitters (i.e., 60-keV gammas from the decay of 241Am and even plutonium L-shell X-rays). 

LLNL can provide a broad spectrum of in vivo counting services with varying degrees of detectability 

(Mansfield 2000).

From at least the late 1980s (and possibly before), action levels were based on the amount of 

radionuclide excreted or detected in the whole body (Mansfield 1989). Later, assessments of intakes 

and even doses were based on in vivo and in vitro monitoring results, using computer programs 

developed by LLNL (Mansfield 2000). Both in vivo and in vitro analysis data records and associated 

interpretations exist from the 1960s and appear to be retrievable.

A review of in-house procedures used to assess the concentration of radioactivity in urine indicates 

that quality control steps were an integral part of the process (LLNL 1979). For example, LLNL ran 

duplicates consistently, and comparisons of results to “known quantities” were a critical step. 

Therefore, in vitro results from in-house processing are generally reliable. However, interpretation of 

those results can be difficult, primarily because they might not have considered the contribution of 

environmental radioactivity (i.e., uranium, thorium). Because sample collection could have occurred 

at work (e.g., “in-field” tritium analyses, Monday morning urines), cross-contamination could be an 

issue.1

Nevertheless, dose reconstructors can prepare reasonably reliable, yet conservative (i.e., designed to 

maximize the resulting dose) estimates of dose for claimants by simply knowing the dates of 

employment, the claimant’s work locations, and in vivo and in vitro bioassay results. Other variables, 

such as particle sizes and clearance classes can be readily reconstructed from historical records. 

This document includes guidance on selecting source terms, interpreting in vivo and in vitro

measurement results, and instructions for assessing dose for monitored and unmonitored claimants. 

The information in this document will be updated as new data, information, and reports become 

available.

 

1

 Twenty-four-hour collections typically took place at the employee’s home on weekends. However, spot samples were 

often collected in site restrooms (Mansfield 1989).
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 8 of 40

5.2 SOURCE TERM

The primary LLNL missions have been weapons research and development, controlled nuclear 

weapons research, peaceful uses of nuclear explosives, biomedical research, and laser fusion 

research. In addition, work in non-nuclear technologies and materials testing has been ongoing since 

at least 1976 (LLNL 1977).

If information on the source term to which the claimant has been exposed is available, dose 

reconstructors should use that source term. However, if no source term information is available, the 

values and parameters in Table 5-1 provide conservative input to the assessment process.

Unless site-specific information is available, the particle size is assumed to be 5-μm activity median 

aerodynamic diameter (AMAD), as recommended in ICRP (1994, paragraph 5). 

If a monitoring result refers to “weapons grade plutonium,” dose reconstructors can use the isotopic 

mix listed in Table 5-2 for material aging times (Mansfield 2000).

If a monitoring result refers to “uranium,” dose reconstructors can use the isotopic mix listed in 

Table 5-3 for enrichment levels (Mansfield 2000).

5.3 MEASUREMENT METHODS

In vivo measurement methods began at the site as far back as 1964, when LLNL used sodium iodide 

detectors connected to simple multichannel analyzers to evaluate 40K and fission products in humans. 

In 1964, a scanning bed, referred to as the “RIDL,” was in use at LLNL. This device was a shadow 

shield counter equipped with a 3-in. by 2-in. sodium iodide detector for detection and quantification of 

relatively high (i.e., greater than 200 keV) gamma photons. It had some capability for locating the 

source of radionuclide deposition in subjects. This device was calibrated with a water tank phantom 

spiked with 137Cs and 40K (Anderson 1964). 

Whole-body counting using a shadow shield counter began in the early 1960s. In April 1965, thin

crystal studies began (LLNL 1965). In 1965 and 1966, LLNL identified 65Zn in people working in 

Building 153, which dominated the study types at that time (Anderson 1964).

LLNL began using dual thin crystals in May 1966 to identify and quantify low-energy photon emitters 

in the chests of workers (Shapiro 1973). In November 1966, Koch hardware that used a 5-in. by 
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 9 of 40

Table 5-1. Source term by building.

a

(Current) building no. Building activity Radionuclideb

Activity

fraction

131 M&E Divisions Th-232 2.29E-10

U-234 8.47E-02

U-235 1.18E-02

U-238 9.04E-01

132 Analytical & Nuclear Chemistry Laboratories; Forensic 

Sciences Center

Th-228 1.2E-10

Th-230 3.5E-10

Th-232 7.9E-05

U-234 3.1E-04

U-235 7.2E-03

U-238 9.9E-01

Pu-238 4.4E-03

Pu-239 6.7E-06

Pu-240 1.7E-06

Pu-241 1.6E-05

Pu-242 2.0E-10

Am-241 3.1E-07

151 Isotope Sciences, Environmental Services Laboratory Ni-63 1.1E-01

I-129 4.6E-10

U-233 1.3E-07

U-234 2.0E-03

U-235 3.0E-04

U-236 7.3E-06

U-238 1.9E-02

Np-237 8.9E-04

Pu-236 5.3E-06

151 Isotope Sciences, Environmental Services Laboratory Pu-238 2.0E-02

Pu-239 1.7E-01

Pu-240 3.8E-02

Pu-241 6.1E-01

Pu-242 9.5E-05

Pu-244 5.1E-05

Am-241 1.0E-02

Am-243 7.6E-04

Cm-242 5.1E-06

Cm-244 1.5E-05

Cm-246 1.2E-05

Cm-248 8.0E-06

Cf-249 1.5E-04

175 U-AVLIS U-234 3.11E-01

U-235 9.50E-03

U-238 6.79E-01

177 U-AVLIS U-234 3.05E-01

U-235 1.04E-02

U-238 6.85E-01

179 U-234 4.38E-01

U-235 1.24E-01

U-238 4.38E-01

212 Physics and Space (rotating target neutron source) H-3 1 

222 H-3 6.96E-01

C-14 5.44E-10

Ni-63 3.04E-01

Th-232 6.86E-08

U-234 2.84E-08

U-235 1.25E-09

U-238 7.13E-06

Pu-239 6.53E-08

223 Pu-238 1.09E-02

Pu-239 8.78E-01

Am-241 1.25E-04

Am-243 1.11E-01

226 H-3 1.00E+00

U-238 9.67E-06
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Table 5-1 (Continued). Source term by building.a

(Current) building no. Building activity Radionuclideb 

Activity

fraction

227 U-234 5.05E-01

U-235 2.26E-02

U-238 4.73E-01

231 Safeguards and engineering Th-232 1.3E-02

U-234 8.5E-02

U-235 1.2E-02

U-238 8.9E-01

235 Characterization studies and ion beam experiments Th-232 4.0E-08

U-234 2.7E-01

U-235 1.6E-02

U-238 6.7E-01

Pu-238 3.0E-04

Pu-239 1.1E-02

Pu-240 2.5E-03

Pu-241 2.9E-02

235 Characterization studies and ion beam experiments Pu-242 1.7E-07

Am-241 1.4E-03

241 R&D C-14 2.1E-07

P-32 8.7E-08

Th-232 1.9E-12

U-234 1.0E+00

U-235 1.8E-06

U-238 4.0E-05

251 Heavy element facility U-233 6.58E-09

Pu-238 2.17E-04

Pu-239 5.54E-09

Pu-243 1.45E-06

Am-241 3.29E-01

Cm-243 1.69E-01

Cm-244 5.02E-01

Cm-248 4.18E-05

Cf-252 1.37E-05

253 Laboratories and counting rooms H-3 1.1E-04

C-14 1.0E-07

P-32 1.0E-06

Sr-90 5.5E-06

Sr-90/Y-90 2.8E-08

Y-90 3.7E-08

Cs-137 1.4E-07

Bi-214 3.3E-01

Pb-214 3.3E-01

Po-218 3.3E-01

Ra-226 1.5E-02

Th-230 2.0E-08

U-234 3.7E-04

U-235 1.6E-05

U-238 3.5E-04

Np-237 4.3E-08

Pu-238 2.6E-06

Pu-239 6.6E-05

Pu-240 1.5E-05

Pu-241 6.5E-04

Pu-242 9.9E-10

Am-241 3.1E-06
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 11 of 40

Table 5-1 (Continued). Source term by building.a

(Current) building no. Building activity Radionuclideb

Activity

fraction

254 Bioassays and analytical services H-3 4.4E-02

C-14 1.4E-02

P-32 4.0E-03

P-33 4.0E-01

S-35 4.1E-02

Sr-90 9.9E-02

Y-90 9.9E-02

I-125 6.3E-02

Po-209 2.6E-05

Th-230 6.0E-03

U-232 2.3E-04

U-233 1.0E-05

U-234 4.1E-04

U-235 1.0E-05

U-236 3.2E-03

U-238 8.9E-07

Pu-238 0.0E+00

Pu-239 9.0E-05

Pu-240 0.0E+00

Pu-241 0.0E+00

Pu-242 5.5E-04

Np-237 4.4E-02

254 Bioassays and analytical services Np-239 1.7E-01

Am-241 3.2E-05

Am-243 9.2E-04

Cm-242 2.5E-04

Cm-244 3.3E-04

Cf-249 2.5E-04

Cf-252 5.4E-03

255 Calibration laboratory H-3 1.00E+00

281 Tracer and dissolution studies H-3 3.7E-02

Be-10 1.5E-02

C-14 2.8E-01

Cl-36 1.5E-02

Ca-41 1.5E-01

Ni-63 3.1E-01

Sr-90 1.5E-02

Th-232 1.1E-06

U-233 1.9E-04

U-234 1.9E-04

U-235 8.1E-06

U-238 1.8E-04

U-233/U-238 1.9E-06

Pu-239 6.3E-04

Pu-242 2.0E-03

Pu-244 2.4E-06

Np-237 1.7E-01

Am-241 5.4E-04

Am-241/Np2371.4E-04

282 Residual contamination H-3 1.0E+00

292 Residual contamination rotating target neutron source H-3 1

298 Laser fusion program H-3 9.9E-01

U-234 4.1E-04

U-235 1.9E-05

U-238 6.0E-03

321 Milling and shaping DU

Thoriumc

321A Milling and shaping U-234 1.8E-01

U-235 2.4E-02

U-238 7.9E-01
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 12 of 40

Table 5-1 (Continued). Source term by building.a

(Current) building no. Building activity Radionuclideb

Activity

fraction

321C Milling, machining, and shaping U-234 6.5E-01

U-235 8.1E-03

U-238 3.4E-01

322 M&E U-234 8.1E-01

U-235 1.2E-02

U-238 1.8E-01

327 M&E U-234 7.7E-02

U-235 8.8E-02

U-238 8.4E-01

331 Research and laboratories H-3 (HTO) 4.69E-01

H-3 (HT) 5.19E-01

H-3 1.24E-02

332 Gloveboxes, HEPA filters Pu-239 TRU

341 Lasers Directorate U-234 5.70E-02

U-235 7.69E-03

U-238 9.35E-01

361 R&D H-3 5.6E-06

C-14 3.3E-03

P-32 8.3E-01

P-33 1.1E-01

S-35 5.4E-02

362 R&D H-3 7.22E-01

C-14 2.78E-01

363 R&D H-3 9.99E-01

C-14 1.11E-07

P-32 1.42E-03

364 R&D H-3 1.11E-03

C-14 2.76E-03

P-32 9.96E-01

365 R&D H-3 4.04E-03

C-14 9.96E-01

366 R&D H-3 5.00E-01

P-32 4.38E-01

P-33 6.25E-02

377 R&D H-3 7.66E-04

C-14 3.83E-03

P-32 9.95E-01

Ni-63 1.15E-04

378 Co-57 1.44E-03

Co-60 5.59E-02

Sr-85 6.52E-02

Cd-109 3.58E-02

Cs-134 6.98E-01

Cs-137 6.98E-02

U-233 2.14E-04

U-234 8.66E-07

U-235 4.84E-08

U-238 2.00E-06

Pu-236 2.13E-03

Pu-239 6.99E-05

Pu-240 5.59E-05

Pu-242 1.05E-02

378 Pu-244 3.58E-04

Np-237 1.44E-06

Am-241 6.00E-02

Am-243 4.01E-04

381 H-3 1

391 H-3 1 
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Table 5-1 (Continued). Source term by building.a

(Current) building no. Building activity Radionuclideb

Activity

fraction

412W Ni-59 3.50E-07

Ni-63 1.00E+00

446 C-14 1

513 Waste processing H-3 9.85E-01

U-233 1.52E-05

U-234 1.08E-03

U-235 1.52E-04

U-238 1.04E-02

514 Waste processing H-3 8.61E-01

514 Waste processing C-14 2.40E-02

P-32 6.53E-02

Th-228 1.47E-05

Th-229 6.04E-05

Th-230 5.81E-05

Th-232 7.87E-05

U-232 6.08E-05

U-233 7.01E-03

U-234 3.60E-03

U-235 2.18E-04

U-236 8.92E-07

U-237 1.82E-09

U-238 8.29E-03

514 Waste processing Pu-236 3.67E-08

Pu-238 4.72E-04

Pu-239 6.74E-03

Pu-240 1.68E-03

Pu-241 1.49E-04

Pu-242 1.60E-04

Pu-244 1.09E-06

Np-237 6.77E-05

Np-239 2.41E-06

Am-241 4.23E-03

Am-243 8.33E-05

Cm-244 6.76E-05

Cf-249 9.30E-06

514 Tank Farm H-3 5.45E-01

P-32 1.01E-01

Sr-90 2.05E-01

Cs-137 1.08E-01

612 Waste storage and repackaging H-3 9.79E-01

612 Waste storage and repackaging U-232 1.58E-05

U-233 2.08E-05

U-234 4.93E-04

U-235 1.85E-04

U-237 3.39E-08

U-238 2.01E-03

612 Yard H-3 1.00E+00

625 Waste operations H-3 2.14E-01

U-234 2.76E-01

U-235 4.89E-01

U-238 2.32E-03

Pu-238 7.68E-05

Pu-239 3.40E-03

Pu-240 3.83E-06

Pu-241 1.30E-02

Pu-242 4.46E-08

Am-241 2.87E-04

2561 U-234 5.00E-01

U-238 5.00E-01

U-234 8.01E-02

U-235 1.12E-02

U-238 8.66E-01

Site 300-801 H-3 1
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 14 of 40

Table 5-1 (Continued). Source term by building.a

(Current) building no. Building activity Radionuclideb

Activity

fraction

Site 300 Pit 7 H-3 1

Site 300-810A U-234 8.38E-02

U-235 1.16E-02

U-238 9.05E-01

Site 300-810B U-234 8.47E-02

U-235 1.13E-02

U-238 9.04E-01

Site 300-850 H-3 9.6E-01

U-234 3.2E-03

U-235 4.4E-04

U-238 3.4E-02

Site 300-851 H-3 9.57E-01

U-234 9.57E-04

U-235 1.34E-04

U-238 1.02E-02

Site 300 Well 8 Spring H-3 1

a. This table was derived from data reported for calendar years 1992 through 2001 as part of the National Emission 

Standards for Hazardous Air Pollutants (NESHAPs; 40 CFR Part 61, Subpart H) reporting process (LLNL 1994, 1995, 

1996, 1997, 1998, 1999, 2000, 2001, 2002, 2003). As part of its required reporting under NESHAPs, LLNL must 

monitor emissions by radionuclide and total activity at all building release points. This table only lists buildings for 

which NESHAPS data was reported; this is not a complete list of all buildings. For a listing of major buildings and 

activities see Lawrence Livermore National Laboratory- Site Description (ORAU 2005.) While the emissions identified 

over the years might not be fully representative of the source term in each building, they do provide a reasonable basis, 

in the absence of claimant-specific monitoring results, for determining the types of radioactive materials to which 

employees might have been exposed. To develop the table, the isotopic listing of each building in the relevant 

NESHAPs reports was captured in a building-specific spreadsheet. The annual emissions were totaled for each 

isotope and then for all isotopes in the listing. Activity fractions were obtained by dividing the activity total by the “all 

isotopes” total (Berger 2005). The Activity fraction for each radionuclide in the table is that fraction of the total source 

term present (i.e., the sum of the fractions for each building is equal to “1" or 100% of the mixture). However, those 

radionuclides with Activity fractions equal to “zero”, those with half-lives too short to contribute significant dose (i.e., 

isotopes of nitrogen and oxygen), and those that contribute submersion dose only were not included in this listing.

b. The list of isotopes was limited to those providing the principal activity or dose. A listing of all isotopes by building is 

included in Attachment A.

c. LLNL (1987a).

Table 5-2. Isotopic information for aging of weapons-grade plutonium. 

Mixture Designation Fresh 5-year 10-year 15-year 20-year 25-year 30-year

Years of Aging 0 5 10 15 20 25 30

Specific Activity in Mixture (Ci/g)

238Pu 6.85E-03 6.59E-03 6.33E-03 6.08E-03 5.85E-03 5.62E-03 5.40E-03

239Pu 5.81E-02 5.81E-02 5.80E-02 5.80E-02 5.80E-02 5.80E-02 5.80E-02

240Pu 1.37E-02 1.37E-02 1.37E-02 1.37E-02 1.37E-02 1.36E-02 1.36E-02

241Pu 5.98E-01 4.70E-01 3.69E-01 2.91E-01 2.28E-01 1.79E-01 1.41E-01

242Pu 1.57E-06 1.57E-06 1.57E-06 1.57E-06 1.57E-06 1.57E-06 1.57E-06

241Am 0.00E+00 4.24E-03 7.54E-03 1.01E-02 1.21E-02 1.36E-02 1.48E-02

239+240Pu 7.18E-02 7.18E-02 7.17E-02 7.17E-02 7.17E-02 7.16E-02 7.16E-02

Pu-alpha 7.87E-02 7.84E-02 7.80E-02 7.78E-02 7.76E-02 7.72E-02 7.70E-02

Activity ratios

239+240Pu:241Am N/A 1.69E+01 9.51E+00 7.10E+00 5.93E+00 5.26E+00 4.84E+00

239+240Pu:238Pu 1.05E+01 1.09E+01 1.13E+01 1.18E+01 1.23E+01 1.27E+01 1.33E+01

Pu alpha: 239+240Pu 1.10E+00 1.09E+00 1.09E+00 1.08E+00 1.08E+00 1.08E+00 1.08E+00

Pu alpha:238Pu 1.15E+01 1.19E+01 1.23E+01 1.28E+01 1.33E+01 1.37E+01 1.43E+01

241Pu: Pu alpha 7.60E+00 6.00E+00 4.73E+00 3.74E+00 2.94E+00 2.32E+00 1.83E+00

1/16th-in. sodium iodide crystal was deployed. Design for the use of hyperpure germanium (HPGe) 

detectors began in August 1991, with deployment presumed shortly thereafter.
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 15 of 40

Table 5-3. Isotopic information for various uranium enrichment levels. 

Mixture Radionuclide

Mass 

fraction

Activity 

fraction

Specific activity 

of mix (Ci/g)

DU U-238 9.975E-1 9.02E-1 3.72E-7

DU U-235 2.50E-3 1.45E-2 3.72E-7

DU U-234 5.00E-6 8.4-E-2 3.72E-7

U-Nat. U-238 9.93E-1 4.77E-1 7.00E-7

U-Nat. U-235 7.20E-3 2.23E-2 7.00E-7

U-Nat. U-234 5.60E-5 5.00E-1 7.00E-7

AVLIS 5% Enriched U-238 9.50E-1 4.11E-1 7.78E-7

AVLIS 5% Enriched U-235 5.00E-2 1.39E-1 7.78E-7

AVLIS 5% Enriched U-234 5.60E-5 4.50E-1 7.78E-7

95% HEU U-238 3.96E-2 1.98E-4 6.27E-5

95% HEU U-235 9.50E-1 3.06E-2 6.27E-5

95% HEU U-234 1.04E-2 9.69E-1 6.27E-5

It is not clear whether whole-body or specific organ counting was performed with this technology for 

routine monitoring over the duration of LLNL history. It appears that few workers were sent for in vivo

bioassay unless there was a workplace trigger (i.e., elevated airborne radionuclide concentrations or a 

spill/release); however, baseline and termination counts occurred as early as 1966 (LLNL 1966).

From at least 1966, in vitro measurement methods were in use, with routine monitoring occurring 

annually and semiannually during those early days. Building 331 had a monthly monitoring program 

in place. During weapons test support periods, routine sampling frequencies for involved employees 

appear to have been increased to quarterly, with weekly sampling for Building 331 (for tritium) and 

monthly sampling for Building 251 (LLNL 1966). Around the mid-1980s, the monitoring frequency 

appears to have been dictated by workplace supervisors (Mansfield 1989, p. 9).

Employees involved in weapons testing at the Nevada Test Site, the Pacific proving grounds, and in 

Colorado were presumably monitored for internal radiation exposures. However, the monitoring 

method, frequency, and detection capabilities have not been determined.

Bioassays were supplemented with workplace and/or personal breathing zone airborne monitoring 

(Mansfield 1989, p. 10). In addition, nasal smears were collected after incidents as a means of 

assessing intake potential. However, data from nasal smears apparently were not used as input to 

dose assessments. Nonetheless, the dose reconstructor might come across some of these data in 

claimant dosimetry records (Mansfield 1989, Appendix A). 

Table 5-4, compiled from a variety of LLNL documents (cited in the table footnote), lists in vivo and in 

vitro sampling frequencies for various employment periods. In the case of in vitro methods, sample 

sizes may have been “spot” (single void), 24-hour collections, or simulated 24-hour collections, with 

creatinine corrections performed if the results were indicative of “significant” intakes (Mansfield 2000).

5.3.1

A variety of program documents reference detection levels for the various in vivo and in vitro bioassay 

methods. These methods are listed in Table 5-5, and the documents are cited in the table footnote.

Measurement Types and Detection Levels
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 16 of 40

Table 5-4. In vivo and in vitro sampling frequencies. 

Routine 

monitoring 

method

Measurement 

type Radionuclide Period Building Frequency

In vitro Urine Pu 2000 - present Site-wide Semiannual

In vivo Lung count Am, Pu 1970 - present Site-wide Annual

In vivo Lung Weapons-grade Pu 1989a Site-Wide Annual

In vitro Urine Weapons-grade Pu 1989a Site-Wide Semiannual

In vitro Urine U 1980a 231 Quarterly

In vitro Urine U 1983a 251 Quarterly

In vitro Urine U 2000 - present Site-wide Quarterly, monthly

In vivo Lung count U 1985a Site 300 Annual

In vitro Urine U 1985a Site 300 Semiannual (collected on Mondays)

In vivo Lung count U, Th 2000 - present Site-wide Annual, semiannual

In vivo Lung count U, Th 1987 Bldg. 321 Annual

In vitro Urine Gross alpha, beta (LSC) 2000 - present Site-wide As applicable

In vivo WB scans MFP 1964 - present

In vivo WB scans 2000 - present Site-wide Annual

In vitro Urine TRU 2000 - present Site-wide Annual, semiannual

In vitro Urine P-32, C-14 (LSC) 2000 - present Site-wide As applicable

In vivo Thyroid count I-131 2000 - present Site-wide Coordinated with work schedule

In vivo Thyroid I-131 1989a Site-Wide Monthly

In vitro Urine Not Specified 1966a 101 Quarterly, annual

In vitro Urine Not Specified 1966a 102 Annual

In vitro Urine Not Specified 1966a 110 Annual

In vitro Urine Not Specified 1966a 112 Annual

In vitro Urine Not Specified 1966a 114C Annual

In vitro Urine Not Specified 1966a 117 Quarterly, annual

In vitro Urine Not Specified 1966a 121 Semiannual

In vitro Urine Not Specified 1966a 125 Semiannual

In vitro Urine Not Specified 1966a 127 Semiannual

In vitro Urine Not Specified 1966a 170 Annual

In vitro Urine Not Specified 1966a 171 Annual

In vitro Urine Not Specified 1966a 172 Weekly, monthly

In vitro Urine Not Specified 1966a 173B Annual

In vitro Urine Not Specified 1966a 190 Semiannual, monthly

In vitro Urine Not Specified 1966a 193 Annual

In vitro Urine P-32 1989a Site-Wide Monthly

In vitro Urine HTO 2000 - present Site-wide Quarterly, monthly, biweekly, or weekly

In vitro Urine HTO 1989a Site-Wide Weekly, monthly

In vitro Urine N-13 1980a 151 Weekly, monthly

In vitro Urine O-15 1980a 151 Weekly, monthly

In vitro Urine Ar-37 1980a 151 Weekly, monthly

In vitro Urine Kr-85 1980a 151 Weekly, monthly

In vivo Wound Pu-239 1961 - present PRN

a. Periods shown as a single year were taken from annual reports which did not indicate a beginning date for the stated bioassay period.

Sources: Mansfield 2000, LLNL 1966 Appendix A; Ozaki 1980; Wilson 1982; Leahy 1983; UC (1961, 1964); Gibson 1985; LLNL 1987b.

In some cases, there is evidence that monitoring by a particular methodology was implemented at 

LLNL, but site-specific detection levels for that methodology were not available. In these cases, 

detection levels for comparable Los Alamos National Laboratory (LANL) methods were used to 

complete Table 5-5 because there is evidence of interlaboratory communications on analytical 

methods over the years, and because both LLNL and LANL were operated by the same contractor at 

those times, thus further increasing the likelihood for shared procedures (Berger 2005).

In a few instances, neither LLNL nor LANL detection levels were available, thus requiring a surrogate 

source of information. The ICRP, in Publication 54, specified nominal detection levels for a variety of 

measurement methods that were comparable to those used at LLNL (ICRP 1988). This industry- 
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 17 of 40

Table 5-5. Bioassay detection levels. Routine 

monitoring 

method

Measurement 

type

Analysis 

method Radionuclide Period MDA

In vitro Urine Aluminum nitrate/NTA * Pu-total alpha 1957-1965 0.03 pCi/24-h sample

In vitro Urine Zinc Sulfate counting * Pu-total alpha 1966 0.03 pCi/24-h sample

In vitro Urine PHA * Pu-239, 238 1967-1987 0.03 pCi/24-h sample

In vitro Urine 24-hr sample, alpha spec “Pure” Pu 2000 - present 0.01 dpm

In vitro Urine 24-hr sample, alpha spec Weapons-grade Pu 2000 - present 0.01 dpm

In vivo Lung Phoswich** Pu-239, Am-241 1970 - present 20, 0.5 nanocuries

In vivo Lung Dual Phoswich Pu-239 1980 - present >16.3 nCi

(2.1 cm CWT)

In vivo Lung HPGe arrays “Pure” Pu-239 2000 - present 200 nCi

In vivo Lung HPGe arrays “Pure” Pu-238 2000 - present 150 nCi

In vivo Lung HPGe arrays (Am-241) Weapons-grade Pu 2000 - present 0.15 nCi

In vitro Urine Chemical extraction with 

proportional counting *

Am-241 1958-1982 0.2 pCi/24-hr

In vitro Urine Co-precipitation and alpha 

spec *

Am-241 1983-1989 0.015 pCi/24-hr sample

In vivo Lung HPGe arrays Am-241 2000 - present 0.15 nCi

In vitro Urine 24-hr sample, alpha spec Am-241 2000 - present 0.01 dpm

In vitro Urine Fluorophotometric * DU, Natural U 1949-1976 4 microgram/liter

In vitro Urine Fluorophotometric * DU, Natural U 1976-1978 1 microgram/liter

In vitro Urine Delayed Neutron activation 

analysis *

U-238/U-235 1978-1982 1 microgram per liter

In vitro Urine Anion exchange * DU, Natural U 1982-1987 1 microgram/50 cm3

In vitro Urine Spot sample KPA DU 1987 - 1999 0.04 μg/L

In vitro Urine Spot sample KPA Natural U 1987 - 1999 0.04 μg/L

In vitro Urine Spot sample KPA AVLIS U 1987 - 1999 0.04 μg/L

In vitro Urine Alpha spectrometry** 95% U 1987 - 1999 0.3 picocurie/liter

In vitro Urine Spot Sample ICP-MS U 2000 - present 0.002 micrograms

In vivo Lung HPGe arrays DU, natural U 2000 - present 1 nCi

In vivo Lung HPGe arrays AVLIS U, Th, U-235 2000 - present 1 nCi

In vivo Lung HPGe arrays 40% U, 95% U 2000 - present 0.1 nCi

In vitro Urine TRU-spec column with 

alpha spectroscopy

Th-232, Th-228 1989-present 0.2 dpm/sample (with Th-232 being 

equivalent to Th-228)

In vivo WB Scan -- Th-232 1970-present 1 nCi (based on Tl-208 with no 

correction for branching ratio)

In vivo WB scan -- Natural Th (Tl-208) 2000 - present 1 nCi

In vitro Urine 24-hr sample, alpha spec Cm-244 2000 - present 0.01 dpm

In vivo Lung HPGe arrays Cm-244 2000 - present 100 nCi

In vitro Urine 24-hr sample, alpha spec Np-237 2000 - present 0.01 dpm

In vivo Lung HPGe arrays Np-237 2000 - present 0.35 nCi

In vivo Lung HPGe arrays Pa-233 2000 - present 1 nCi

In vitro Urine Extraction and gross alpha* Po-210 1954-1999 0.1 picocurie/liter

In vitro Urine 24-hr sample, alpha spec Po-210 2000 - present 0.01 dpm

In vitro Urine GM counting * HTO 1952-1958 1 microcurie/liter

In vitro Urine Internal GM counting * HTO 1958-1968 1 microcurie/liter

In vitro Urine Liquid scintillation * HTO 1969-2000 0.02 microcurie/literIn vitro Urine Liquid scintillation HTO 2000 - present 0.01 μCi/L

In vitro Urine Gross beta * Sr-90 1960-1999 See Sr-90 by in-vivo 

bremsstrahlung for 1970 to 1999; 

rest 1 picocurie/liter

In vitro Urine Gross beta and chemical 

extraction**

Sr-90 2000 - present 11 picocuries/liter

In vitro Urine Gross beta ** C-14 1966-1999 108 picocuries per liter

In vitro Urine Liquid scintillation C-14 2000 - present 0.002 μCi/L

In vitro Urine Gross Beta ** I-125, I-131, S-35 1960-1999 100 picocuries per liter

In vitro Urine Liquid scintillation I-125, I-131, S-35 2000 - present 0.002 μCi/L

In vivo Thyroid count -- I-131 2000 - present 0.01 nCi

In vivo WB scan -- I-131 (TB) 1970 - present 1 nCi

In vivo Thyroid count -- I-125 2000 - present 0.02 nCi

In vitro Urine Not specified * P-32 <1975-1983 40 picocuries/L
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 18 of 40

Table 5-5 (Continued). Bioassay detection levels. Routine 

monitoring 

method

Measurement 

type

Analysis 

method Radionuclide Period MDA

In vitro Urine Liquid scintillation P-32, P-33 2000 - present 0.002 μCi/L

In vivo WB scan Shadow shield counter 150- to 300-keV 

gamma emitters

1964 - present 500 gammas per second between 

150 and 300 keV

In vivo WB scan Shadow shield counter K-40 1964 – present 150 g K in 20 min (~120 nCi K-40)

In vivo WB scan -- Cs-137 1970 - present 1 nCi

In vivo WB scan -- Co-60 1970 - present 1 nCi

In vivo WB scan -- Ce-144 1970 - present 8 nCi

In vivo WB scan -- Ce-141 1970 - present 2 nCi

In vivo WB scan -- Ru-103 1970 - present 3 nCi

In vivo WB scan -- Ru-106 1970 - present 5 nCi

In vivo WB scan -- Zr-95 1970 - present 1 nCi

In vivo WB scan -- Mn-54 1970 - present 1 nCi

In vivo WB scan -- Zn-65 1970 - present 1 nCi

In vivo WB scan -- Na-22 1970 - present 1 nCi

Sources: Mansfield 2000; UC 1964; LRL 1970; Griffith 1980; Berger 2005; Argall 2004..

* Based on LANL information

** Based on ICRP 54 information

standard and peer-reviewed source of information was used to complete Table 5-5 pending future 

discovery of either LLNL- or LANL-specific replacement information.

The expected intake pattern in most cases is acute intake. At LLNL, airborne and surface 

contamination was typically controlled to prevent intakes so most intakes would have been the result 

of unexpected releases. Small intermittent releases could have occurred that were not immediately 

detectable, so an individual could have had multiple acute intakes. 

5.3.2

A variety of codes occur on urine bioassay records generated at or for the LLNL. Table 5-6 is a 

summary of the codes known at the present time, along with their interpretation. 

Reporting Formats and Codes

Table 5-6. Bioassay record codes. 

Column no. Type of entry Acceptable entries

1 - 6 Employee number 6-digit integer

8 - 27 Employee name Last name, comma, other names and/or initials

28 - 32 Nuclide analyzed MFP, ALPHA, PU239, PU238, U238, AM241, CM244, or other nuclides 

(maximum of 5 characters)

34 - 39 Date Month, day, year (2 digits each)

41 - 48 Results Floating point number between 9999999. and 0.000 (F8.3 format)

50 - 51 Error One standard deviation as a percentage of the result

53 Sample type U = urine, B = blood, f = feces, n = nose swipe

55 Sample type R = routine; S = special

57 - 60 Address Building or trailer number, e.g., B253 or 2532

61 - 63 L code 3-digit L code

65 Frequency W = weekly; M = monthly; Q = quarterly; S-semiannually; A = annually

67 - 70 Location of exposure Building number, e.g., B253

72 - 76 Analysis number Integer value of analysis number assigned to given sample by Bioassay 

Laboratory

79 - 80 Method Analytical method used to process the samples: 1 = gross alpha 

technique (since 1957); 2 = gross beta technique (since 1974); 3 = 

plutonium separation technique (since 1963); 4 = fluorometric uranium 

technique (since 1974); 5 – 99 = reserved for future revised techniques
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 19 of 40

The practice of offsite collection of samples, which takes place approximately 24 to 48 hours after 

leaving the site, not only minimizes the possibility of sample cross-contamination, but also ensures 

sample collection after the transfer of the rapid clearance component. Some LLNL employees might 

have been asked to submit samples after 1 or 2 days off from work; there could be notation of that 

instruction on the analytical record.

LLNL typically collected urine samples in the workplace, usually on a Wednesday. Therefore, 

contamination of samples from the worker’s hands or clothing cannot be ruled out as a contributor to 

any given result. If a second analysis was performed and if that result was “negative,” sample crosscontamination could have been likely during the first.

For in vitro uranium bioassays, a mean background excretion rate for area residents of 0.019 +/- 

0.047 (one standard deviation) micrograms per day (males and females) may be used (Mansfield

2000). 

If a data set shows an unusually high urinalysis result for a given radionuclide, and if follow-up 

samples were not consistent with the high result, dose reconstructors can consider the high result an 

outlier and disregard it. However, if the result is not obviously an outlier, it is claimant-favorable to 

assume the result is real.

Uncertainties associated with bioassay measurements were not stated in the records.

5.3.3

The in vivo bioassay sampling program at LLNL was extensive, and there was clear direction to 

supervisors on individual employee participation in the program. However, there is evidence that 

participation was not always enforced. Therefore, a means of assessing intake for unmonitored 

claimants is necessary. There is little correlation between external dosimetry and the bioassay 

program. From the mid-1950’s external dosimetry was administered as a site-wide requirement 

whereas bioassay was determined by facility supervisors.

Assessment of Intake for Unmonitored Claimants

For monitored claimants, the omission of regularly scheduled bioassays from the data set will, if 

interpreted as described in Internal Dose Reconstruction Implementation Guide (NIOSH, 2002), result 

in claimant-favorable assessments of intake. 

5.4 SIGNIFICANT INCIDENTS WITH INTERNAL DOSE POTENTIAL 

During operations at LLNL, a number of incidents increased the potential for intakes of radioactive 

materials. If a claimant recalls involvement in one or more of those incidents, dose reconstructors can 

use the information in Table 5-7 as input to an incident-specific assessment. This list is not all 

encompassing, many other incidents probably occurred; these are the incidents identified from review 

of the data capture records. Individual claimant records may provide documentation of involvement in 

other incidents.
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 20 of 40

Table 5-7. Input parameters for significant incidents. 

Incident date Incident description Facility Other information

March 26, 1963 Criticality Bldg. 110 Potential for internal and external exposure 

(iodine, krypton, xenon); monitored participants 

listed by name in 1963 reference (see below)

November 6, 1975 Radium contamination Bldg. 343, Rm. 1005 Crystal was removed from radium dial, resulting 

in spread of contamination

November 23, 1971 Plutonium-contaminated 

squib valve

Bldg. 343 Direct and indirect bioassays performed on 14 

individuals

Sources: Unknown 1963a,b; Leahy 1975; AEC 1971.
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 21 of 40

REFERENCES

AEC (U.S. Atomic Energy Commission), 1971, Investigation Committee Report on Contamination 

Incident at Bldg. 343., Final Draft, San Francisco Operations Office, San Francisco, California.

Anderson, A. L., 1964, Excerpts from Larry Anderson Logbook, Lawrence Livermore Laboratory, 

Livermore, California.

Argall, R. S., 2004, “Technical Basis Document for the Los Alamos National Laboratory – 

Occupational Internal Dose”, ORAUT-TKBS-0010-5, December 21, 2004, Cincinnati, Ohio. 

Berger, C. D., 2005, “Technical Basis Issue in ORAUT-TKBS-0035-5”, memorandum to file, Oak 

Ridge Associated Universities, Oak Ridge, Tennessee.

Gibson, T., 1985, Discipline Action Plan; Radiation Safety Program for Site 300, Lawrence Livermore 

National Laboratory, Livermore, California, April.

Griffith, R. V., 1980, Hazards Control Progress Report 1980, UCRL-50007-80, Lawrence Livermore 

National Laboratory, Livermore, California, June 30.

ICRP (International Commission on Radiological Protection), 1974, Reference Man, Publication 23, 

Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1988, Individual Monitoring for Intakes of 

Radionuclides by Workers: Design and Interpretation, Publication 54, Pergamon Press, 

Oxford, England.

ICRP (International Commission on Radiological Protection), 1994, Dose Coefficients for Intakes of 

Radionuclides by Workers, Publication 68, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1997, Individual Monitoring for Internal 

Exposure of Workers, Publication 78, Pergamon Press, Oxford, England.

Leahy, E. J., 1975, Radium Contamination Incident, Bldg. 343, High Pressure Laboratory, LLNL 

Incident Report No. 0132, Lawrence Livermore National Laboratory, Livermore, California.

Leahy, E. J., 1983, Bioassay Uranium Analysis Program, memorandum to Health Physicists, 

Lawrence Livermore National Laboratory, Livermore, California, December 12.

LLNL (Lawrence Livermore National Laboratory), 1965, Hazards Control Quarterly Report No. 22, 

UCRL-14534, July-September, pg. 13, “Developments at the Whole Body Counter Facility”, 

Lawrence Livermore Laboratory, Livermore, California.

LLNL (Lawrence Livermore National Laboratory), 1966, Hazards Control Manual, Procedure 1003, 

Livermore, California.

LLNL (Lawrence Livermore National Laboratory), 1977, Environmental Monitoring at the Lawrence 

Livermore Laboratory - 1977 Report, LLNL Report No. UCRL-50027-76, Livermore, California, 

March 11.

LLNL (Lawrence Livermore National Laboratory), 1979, Radiochemical Procedures for Bioassay, 

Report No. M-104, Livermore, California.
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 22 of 40

LLNL (Lawrence Livermore National Laboratory), 1987a, Radiation Safety Discipline Action Plan, 

Buildings 341, 343, and 345, Livermore California, May.

LLNL (Lawrence Livermore National Laboratory), 1987b, Radiation Safety Discipline Action Plan, 

Building 321, Livermore California, May.

LLNL (Lawrence Livermore National Laboratory), 1994, LLNL NESHAPs 1993 Annual Report, UCRLID-113867-94, Environmental Monitoring and Analysis Division, Environmental Protection 

Department, Livermore, California.

LLNL (Lawrence Livermore National Laboratory), 1995, LLNL NESHAPs Project 1994 Annual Report

UCRL-ID-113867-95, Environmental Monitoring and Analysis Division, Environmental 

Protection Department, Livermore, California.

LLNL (Lawrence Livermore National Laboratory), 1996, LLNL NESHAPs Project 1995 Annual Report, 

UCRL-ID-113867-96, Environmental Monitoring and Analysis Division, Environmental 

Protection Department, Livermore, California.

LLNL (Lawrence Livermore National Laboratory), 1997, LLNL NESHAPs Project 1996 Annual Report

UCRL-ID-113867-97, Environmental Monitoring and Analysis Division, Environmental 

Protection Department, Livermore, California.

LLNL (Lawrence Livermore National Laboratory), 1998, LLNL NESHAPs 1997 Annual Report, UCRLID-113867-98, Environmental Monitoring and Analysis Division, Environmental Protection 

Department, Livermore, California.

LLNL (Lawrence Livermore National Laboratory), 1999, LLNL NESHAPs 1998 Annual Report, UCRLID-113867-99, Environmental Monitoring and Analysis Division, Environmental Protection 

Department, Livermore, California.

LLNL (Lawrence Livermore National Laboratory), 2000, LLNL NESHAPs 1999 Annual Report, UCRLID-113867-00, Environmental Monitoring and Analysis Division, Environmental Protection 

Department, Livermore, California.

LLNL (Lawrence Livermore National Laboratory), 2001, LLNL NESHAPs 2000 Annual Report, UCRLID-113867-01, Environmental Monitoring and Analysis Division, Environmental Protection 

Department, Livermore, California.

LLNL (Lawrence Livermore National Laboratory), 2002, LLNL NESHAPs 2001 Annual Report, UCRLID-113867-02, Environmental Monitoring and Analysis Division, Environmental Protection 

Department, Livermore, California.

LLNL (Lawrence Livermore National Laboratory), 2003, LLNL NESHAPs 2002 Annual Report, UCRLID-113867-03, Environmental Monitoring and Analysis Division, Environmental Protection 

Department, Livermore, California.

LRL (Lawrence Radiation Laboratory), 1970, Hazards Control Progress Report No. 36, January 

Through May 1970, UCRL-5007-70-1, Livermore California.

Mansfield, W. G., 1989, Health and Safety Manual - LLNL Internal Dosimetry Program Manual, 

Supplement 33.10, Livermore, California.
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 23 of 40

Mansfield, W. G. and A. R. Wood-Zika, 1989, Technical Basis Manual for Internal Dosimetry, UCRLMA-147534, Lawrence Livermore National Laboratory, Livermore, California.

NCRP (National Council on Radiation Protection and Measurements), 1987, Use of Bioassay 

Procedures for Assessment of Internal Radionuclide Deposition, Report No. 87, Bethesda, 

Maryland, February.

NIOSH (National Institute for Occupational Safety and Health), 2002, Internal Dose Reconstruction 

Implementation Guide, OCAS-IG-002, Office of Compensation Analysis and Support, 

Cincinnati, Ohio.

ORAU(Oak Ridge Associated Universities), 2005, Technical Basis Document for Lawrence Livermore 

National Laboratory - Site Description, ORAUT-TKBS-0035-2, Oak Ridge, Tennessee. 

NRC, 1974, U. S. Nuclear Regulatory Commission, Regulatory Guide No. 8.11, Applications of 

Bioassay for Uranium, June, 1974, Washington, DC.

Ozaki, C. B., 1980, Estimation of Total Absorbed Dose from Inhalation of N-13, O-15, Ar-37 and Kr85, memorandum to Paul Meyer, Lawrence Livermore National Laboratory, Livermore, 

California, May 23.

Potter, C. A., 2002, Intake Retention Fractions Developed from Models Used in the Determination of 

Dose Coefficients Developed for ICRP Publication 68 - Particulate Inhalation, Health Physics, 

Vol. 83, No. 5, November.

Shapiro, E. G. and A. L. Anderson, 1973, Dual Energy Analysis Using Phoswich Scintillation 

Detectors for Low-Level In-vivo Counting, UCRL-74797, Livermore, California, October 3, 

1973.

UC (University of California), 1961, Hazards Control Quarterly Report No. 5 (April-June, 1961), UCRL6567, Livermore, California, July.

UC (University of California), 1964, Hazards Control Quarterly Report No. 17 (April - June 1964), 

UCRL-12004, Livermore, California.

Unknown, 1963a, Criticality Event of 1963, data capture file identifier CritEvent03261963, Lawrence 

Livermore National Laboratory, Livermore, California.

Unknown, 1963b, Criticality Event of 3/26/1963 - Radiation Exposure, Contamination Levels, and Air 

Activities (excerpt from unnamed report), data capture identifier 

CritEvent03261963RadExposures, Lawrence Livermore National Laboratory, Livermore, 

California.

Wilson, R. L., Establishing Criteria for Routine Uranium Bioassay in Building 231, memorandum to 

Distribution, Lawrence Livermore National Laboratory, Livermore, California, November 2.
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 24 of 40

GLOSSARY

activity median aerodynamic diameter (AMAD)

The diameter of a unit density sphere with the same terminal settling velocity in air as that of 

the aerosol particle whose activity is the median for the entire aerosol.

acute

Pertaining to intakes received “acutely,” i.e., within a short period. 

bioassay

Measurement of amount or concentration of radioactive material either in the body or in 

biological material excreted or removed from the body. Another word for radiobioassay. 

bioassay procedure

A procedure used to determine the kind, quantity, location, and retention of radionuclides in 

the body by direct (in vivo) measurements or by in vitro analysis of material excreted or 

removed from the body.

body burden

The quantity of radioactive material contained in the individual's body at a particular point in 

time.

chronic

Pertaining to low-level intakes received on a prolonged basis.

dose 

A general term for absorbed dose, dose equivalent, effective dose equivalent, committed dose 

equivalent, committed effective dose equivalent, or total effective dose equivalent.

dose equivalent (H)

The product of absorbed dose (D) in rad (or gray) in tissue, a quality factor (Q), and other 

modifying factors (N). Dose equivalent is expressed in units of rem (or sievert) (1 rem = 0.01 

sievert).

exposure

The general condition of being subjected to ionizing radiation, such as by exposure to ionizing 

radiation from external sources or to ionizing radiation sources inside the body. In this 

document, exposure does not refer to the radiological physics concept of charge liberated per 

unit mass of air.

intake

The amount of radionuclide taken into the body by inhalation, absorption through intact skin, 

injection, ingestion, or through wounds. Depending on the radionuclide involved, intakes can 

be reported in units of mass, activity, or potential alpha energy.

internal dose or exposure

The dose equivalent received from radioactive material taken into the body (i.e., internal 

sources).
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 25 of 40

internal dose assessment

An assessment of the intake and associated internal radiation dose to workers based on 

measurements taken in the work environment or from individual bioassay measurements.

in vitro measurement

Measurements to determine the presence of or to estimate the amount of radioactive material 

in the excreta or in other biological materials removed from the body.

in vivo measurement

The measurement of radioactive material in the human body utilizing instrumentation that 

detects radiation emitted from the radioactive material in the body.

lung solubility type (F, M, or S)

A classification scheme for inhaled material according to its rate of clearance from the 

pulmonary region of the lung.

minimum detectable amount (MDA)

The smallest amount (activity or mass) of an analyte in a sample that will be detected with a 

probability of nondetection (Type II error) while accepting a probability of erroneously deciding 

that a positive (non-zero) quantity of analyte is present in an appropriate blank sample (Type I 

error).

minimum detectable concentration (MDC)

The minimum detectable amount, MDA, expressed in units of concentration.

monitoring (personnel)

The measurement of radioactivity in the whole body, in a region of the body, in material 

eliminated from the body or in the air for reasons related to the estimation of intake of 

radioactive material. The term monitoring includes interpretation of the measurements.

occupational dose

An individual's ionizing radiation dose (external and internal) resulting from that individual's 

work assignment. Occupational dose does not include doses received as a medical patient or 

doses resulting from background radiation or participation as a subject in medical research 

programs.

radiation

Ionizing radiation: alpha particles, beta particles, gamma rays, X-rays, neutrons, high-speed 

electrons, high-speed protons, and other particles capable of producing ions. Radiation, as 

used in this document, does not include nonionizing radiation, such as radio- or microwaves, 

or visible, infrared, or ultraviolet light.

rem

A special unit for dose equivalent. One rem is equal to 0.01 sievert.

routine monitoring

Monitoring carried out at regular intervals during normal operations.

sievert

The special name for the International System unit of dose equivalent. One sievert equals 1 

joule per kilogram, which equals 100 rem. 
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 26 of 40

special monitoring

Monitoring carried out in actual or suspected abnormal conditions (i.e., measurements 

performed to estimate the amount of radionuclide deposited in a person when an intake is 

known or is suspected to have occurred). 

spot sample

A single void of urine.
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 27 of 40

ATTACHMENT A

SOURCE TERM BY BUILDING

(Page 1 of 14)

Table A-1. Source term by building.

a

(Current) building no. Building activity Radionuclide

Activity

fraction

131 M&E Divisions Th-232 2.29E-10

U-234 8.47E-02

U-235 1.18E-02

U-238 9.04E-01

132 Analytical & Nuclear Chemistry Laboratories; Forensic 

Sciences Center

H-3 7.3E-09

Co-60 1.5E-07

Th-228 1.2E-10

Th-230 3.5E-10

Th-232 7.9E-05

U-234 3.1E-04

U-235 7.2E-03

U-238 9.9E-01

Pu-238 4.4E-03

Pu-239 6.7E-06

Pu-240 1.7E-06

Pu-241 1.6E-05

Pu-242 2.0E-10

Am-241 3.1E-07

Ni-63 1.1E-01

Zn-65 5.1E-07

Sr-90 3.8E-05

Y-90 3.7E-06

Tc-99 1.3E-10

Tc-99m 5.3E-08

Ru-106 4.6E-09

Sn-113 1.4E-07

Sb-125 1.6E-06

I-129 4.6E-10

Ba-133 5.1E-08

Cs-134 5.1E-08

Cs-136 7.1E-07

Cs-137 9.5E-05

Ce-144 8.0E-07

Pm-147 7.6E-05

Eu-152 6.5E-08

Eu-154 6.5E-08

Eu-155 1.6E-12

Bi-207 4.2E-05

Po-209 1.9E-12

Ra-226 2.5E-08

Pa-231 2.7E-07

Th-228 1.4E-09

Th-229 2.2E-12

Th-230 1.5E-09

Th-232 6.4E-08

U-232 2.7E-11

U-233 1.3E-07

U-234 2.0E-03

U-235 3.0E-04

U-236 7.3E-06

U-238 1.9E-02

Np-237 8.9E-04

Pu-236 5.3E-06
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 28 of 40

ATTACHMENT A

SOURCE TERM BY BUILDING

(Page 2 of 14)

Table A-1 (Continued). Source term by building.a

(Current) building no. Building activity Radionuclide

Activity

fraction

151 Isotope Sciences, Environmental Services Laboratory Pu-238 2.0E-02

Pu-239 1.7E-01

Pu-240 3.8E-02

Pu-241 6.1E-01

Pu-242 9.5E-05

Pu-244 5.1E-05

Am-241 1.0E-02

Am-243 7.6E-04

Cm-242 5.1E-06

Cm-244 1.5E-05

Cm-246 1.2E-05

Cm-248 8.0E-06

Cf-249 1.5E-04

175 U-AVLIS U-234 3.11E-01

U-235 9.50E-03

U-238 6.79E-01

177 U-AVLIS U-234 3.05E-01

U-235 1.04E-02

U-238 6.85E-01

179 U-234 4.38E-01

U-235 1.24E-01

U-238 4.38E-01

212 Physics and Space (rotating target neutron source) H-3 1

222 H-3 6.96E-01

C-14 5.44E-10

Ni-63 3.04E-01

Th-232 6.86E-08

U-234 2.84E-08

U-235 1.25E-09

U-238 7.13E-06

Pu-239 6.53E-08

223 Pu-238 1.09E-02

Pu-239 8.78E-01

Am-241 1.25E-04

Am-243 1.11E-01

226 H-3 1.00E+00

U-238 9.67E-06

227 U-234 5.05E-01

U-235 2.26E-02

U-238 4.73E-01

231 Safeguards and engineering Th-232 1.3E-02

U-234 8.5E-02

U-235 1.2E-02

U-238 8.9E-01

235 Characterization studies and ion beam experiments Th-232 4.0E-08

U-234 2.7E-01

U-235 1.6E-02

U-238 6.7E-01

Pu-238 3.0E-04

Pu-239 1.1E-02

Pu-240 2.5E-03

Pu-241 2.9E-02
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 29 of 40

ATTACHMENT A

SOURCE TERM BY BUILDING

(Page 3 of 14)

Table A-1 (Continued). Source term by building.a

(Current) building no. Building activity Radionuclide

Activity

fraction

235 Characterization studies and ion beam experiments Pu-242 1.7E-07

Am-241 1.4E-03

241 R&D C-14 2.1E-07

P-32 8.7E-08

Th-232 1.9E-12

U-234 1.0E+00

U-235 1.8E-06

U-238 4.0E-05

251 Heavy element facility U-233 6.58E-09

Pu-238 2.17E-04

Pu-239 5.54E-09

Pu-243 1.45E-06

Am-241 3.29E-01

Cm-243 1.69E-01

Cm-244 5.02E-01

Cm-248 4.18E-05

Cf-252 1.37E-05

253 Laboratories and counting rooms H-3 1.1E-04

C-14 1.0E-07

P-32 1.0E-06

Sr-90 5.5E-06

Sr-90/Y-90 2.8E-08

Y-90 3.7E-08

Cs-137 1.4E-07

Bi-214 3.3E-01

Pb-214 3.3E-01

Po-218 3.3E-01

Ra-226 1.5E-02

Th-230 2.0E-08

U-234 3.7E-04

U-235 1.6E-05

U-238 3.5E-04

Np-237 4.3E-08

Pu-238 2.6E-06

Pu-239 6.6E-05

Pu-240 1.5E-05

Pu-241 6.5E-04

Pu-242 9.9E-10

Am-241 3.1E-06

254 Bioassays and analytical services H-3 4.4E-02

C-14 1.4E-02

P-32 4.0E-03

P-33 4.0E-01

S-35 4.1E-02

Sr-90 9.9E-02

Y-90 9.9E-02

I-125 6.3E-02

Po-209 2.6E-05

Th-230 6.0E-03

U-232 2.3E-04

U-233 1.0E-05

U-234 4.1E-04

U-235 1.0E-05

U-236 3.2E-03

U-238 8.9E-07

Pu-238 0.0E+00

Pu-239 9.0E-05
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 30 of 40

ATTACHMENT A

SOURCE TERM BY BUILDING

(Page 4 of 14)

Table A-1 (Continued). Source term by building.a

(Current) building no. Building activity Radionuclide

Activity

fraction

254 Bioassays and analytical services Pu-240 0.0E+00

Pu-241 0.0E+00

Pu-242 5.5E-04

Np-237 4.4E-02

Np-239 1.7E-01

Am-241 3.2E-05

Am-243 9.2E-04

Cm-242 2.5E-04

Cm-244 3.3E-04

Cf-249 2.5E-04

Cf-252 5.4E-03

255 Calibration laboratory H-3 1.00E+00

C-14 3.50E-10

P-32 2.80E-08

S-35 1.20E-09

Sr-90 4.89E-13

Y-90 4.89E-13

I-125 5.91E-08

I-131 1.80E-07

Th-230 9.53E-11

Th-232 2.50E-15

U-233 2.50E-10

U-234 1.10E-11

U-235 2.90E-13

U-236 8.69E-09

U-238 4.19E-14

Pu-239 2.77E-12

Pu-242 1.40E-11

Np-237 1.70E-09

Np-239 1.70E-09

Am-241 9.54E-13

Am-243 2.54E-11

Cm-242 9.09E-12

Cm-244 9.49E-13

Cf-252 8.39E-11

281 Tracer and dissolution studies H-3 3.7E-02

Be-10 1.5E-02

C-14 2.8E-01

Na-22 1.2E-04

Cl-36 1.5E-02

Ca-41 1.5E-01

Mn-54 4.9E-06

Ni-59 1.0E-04

Ni-63 3.1E-01

Co-60 5.0E-06

Sr-90 1.5E-02

Tc-99 1.5E-04

Sb-125 4.9E-08

Eu-152/Tm-171 1.7E-08

Eu-154/Eu-155 2.1E-05

Th-232 1.1E-06

U-233 1.9E-04

U-234 1.9E-04

U-235 8.1E-06

U-238 1.8E-04

U-233/U-238 1.9E-06

Pu-239 6.3E-04
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 31 of 40

ATTACHMENT A

SOURCE TERM BY BUILDING

(Page 5 of 14)

Table A-1 (Continued). Source term by building.a

(Current) building no. Building activity Radionuclide

Activity

fraction

Pu-242 2.0E-03

Pu-244 2.4E-06

Np-237 1.7E-01

Am-241 5.4E-04

Am-241/Np2371.4E-04

282 Residual contamination H-3 1.0E+00

Rb-86/87 7.5E-09

292 Residual contamination rotating target neutron source H-3 1

298 Laser fusion program H-3 9.9E-01

U-234 4.1E-04

U-235 1.9E-05

U-238 6.0E-03

321 Milling and shaping DU

Thoriumb

321A Milling and shaping U-234 1.8E-01

U-235 2.4E-02

U-238 7.9E-01

321C Milling, machining, and shaping U-234 6.5E-01

U-235 8.1E-03

U-238 3.4E-01

322 M&E U-234 8.1E-01

U-235 1.2E-02

U-238 1.8E-01

327 M&E U-234 7.7E-02

U-235 8.8E-02

U-238 8.4E-01

331 Research and laboratories H-3 (HTO) 4.69E-01

H-3 (HT) 5.19E-01

H-3 1.24E-02

332 Gloveboxes, HEPA filters Pu-239 TRU

341 Lasers Directorate U-234 5.70E-02

U-235 7.69E-03

U-238 9.35E-01

361 R&D H-3 5.6E-06

C-14 3.3E-03

P-32 8.3E-01

P-33 1.1E-01

S-35 5.4E-02

362 R&D H-3 7.22E-01

C-14 2.78E-01

363 R&D H-3 9.99E-01

C-14 1.11E-07

P-32 1.42E-03

364 R&D H-3 1.11E-03

C-14 2.76E-03

P-32 9.96E-01

365 R&D H-3 4.04E-03

C-14 9.96E-01

366 R&D H-3 5.00E-01

P-32 4.38E-01

P-33 6.25E-02

377 R&D H-3 7.66E-04

C-14 3.83E-03

P-32 9.95E-01

Ni-63 1.15E-04

378 Co-57 1.44E-03
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 32 of 40

ATTACHMENT A

SOURCE TERM BY BUILDING

(Page 6 of 14)

Table A-1 (Continued). Source term by building.a

(Current) building no. Building activity Radionuclide

Activity

fraction

378 Co-60 5.59E-02

Sr-85 6.52E-02

Cd-109 3.58E-02

Cs-134 6.98E-01

Cs-137 6.98E-02

U-233 2.14E-04

U-234 8.66E-07

U-235 4.84E-08

U-238 2.00E-06

Pu-236 2.13E-03

Pu-239 6.99E-05

Pu-240 5.59E-05

Pu-242 1.05E-02

Pu-244 3.58E-04

Np-237 1.44E-06

Am-241 6.00E-02

Am-243 4.01E-04

381 H-3 1 

391 H-3 1

412W Ni-59 3.50E-07

Ni-63 1.00E+00

446 C-14 1

513 Waste processing H-3 9.85E-01

C-14 2.49E-04

P-32 3.17E-06

K-40 5.79E-04

Mn-54 8.34E-06

Co-57 7.28E-07

Co-60 1.03E-04

Sr-90 7.27E-04

Nb-95 8.34E-06

Zr-95 1.30E-05

Ru-106 6.07E-09

I-125 7.49E-07

I-131 2.70E-08

Ba-133 6.09E-07

Cs-134 4.13E-06

Cs-137 2.61E-05

Cs-138 1.52E-10

Ce-141 6.07E-09

Ce-144 9.56E-06

Eu-152 1.67E-04

Eu-154 2.12E-07

Eu-155 3.79E-08

Tl-208 6.22E-08

Bi-212 8.50E-08

Bi-214 5.92E-09

Pb-210 7.89E-06

Pb-212 1.43E-07

Pb-214 2.12E-08

Ra-223 4.55E-07

Ra-226 8.65E-08

Ra-228 4.55E-16

Ac-228 1.52E-09

Pa-231 8.50E-07

Th-226 9.71E-07

Th-227 1.67E-06
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 33 of 40

ATTACHMENT A

SOURCE TERM BY BUILDING

(Page 7 of 14)

Table A-1 (Continued). Source term by building.a

(Current) building no. Building activity Radionuclide

Activity

fraction

514 Th-228 4.04E-05

Th-232 3.51E-06

Th-234 3.03E-06

U-233 1.52E-05

U-234 1.08E-03

U-235 1.52E-04

U-238 1.04E-02

Pu-238 1.74E-09

Pu-239 2.28E-04

Pu-240 2.03E-08

Pu-241 1.27E-06

Pu-242 6.87E-07

Am-241 7.62E-04

Cm-244 3.03E-05

Cf-249 2.28E-07

H-3 8.61E-01

Be-7 7.53E-06

C-14 2.40E-02

Na-22 9.05E-05

P-32 6.53E-02

S-35 8.42E-03

K-40 6.11E-05

Sc-46 5.46E-06

Cr-51 5.59E-06

Fe-55 5.42E-06

Mn-54 1.66E-05

Co-56 1.82E-07

Co-57 7.56E-05

Co-58 5.46E-06

Co-60 1.09E-04

Ni-63 1.45E-07

Zn-65 1.92E-08

Y-88 5.33E-05

Sr-89 4.96E-11

Sr-90 9.13E-04

Nb-94 1.19E-06

Nb-95 8.46E-06

Zr-95 1.20E-07

Tc-99 6.05E-05

Ru-103 2.18E-09

Ru-106 1.33E-06

Cd-109 2.89E-08

Sb-125 2.08E-05

I-125 1.39E-04

I-131 3.63E-06

Ba-133 1.02E-04

Cs-134 6.37E-05

Cs-137 1.16E-03

Ce-139 7.14E-13

Ce-141 1.82E-09

Ce-144 4.87E-03

Gd-148 6.04E-05

Pm-147 3.07E-06

Sm-151 2.45E-07

Eu-152 2.29E-04

Eu-154 2.32E-04

Eu-155 2.54E-05
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 34 of 40

ATTACHMENT A

SOURCE TERM BY BUILDING

(Page 8 of 14)

Table A-1 (Continued). Source term by building.a

(Current) building no. Building activity Radionuclide

Activity

fraction

514 Waste processing Hf-172 7.33E-06

Lu-173 1.92E-06

Lu-174 5.42E-06

W-185 5.62E-09

Po-209 9.30E-06

Po-210 9.30E-06

Bi-207 4.71E-06

Bi-210 7.14E-07

Pb-210 5.62E-05

Ra-226 9.73E-06

Th-228 1.47E-05

Th-229 6.04E-05

Th-230 5.81E-05

Th-232 7.87E-05

U-232 6.08E-05

U-233 7.01E-03

U-234 3.60E-03

U-235 2.18E-04

U-236 8.92E-07

U-237 1.82E-09

U-238 8.29E-03

Pu-236 3.67E-08

Pu-238 4.72E-04

Pu-239 6.74E-03

Pu-240 1.68E-03

Pu-241 1.49E-04

Pu-242 1.60E-04

Pu-244 1.09E-06

Np-237 6.77E-05

Np-239 2.41E-06

Am-241 4.23E-03

Am-243 8.33E-05

Cm-244 6.76E-05

Cf-249 9.30E-06

514 Tank Farm H-3 5.45E-01

Be-7 3.33E-07

C-14 2.59E-03

Na-22 5.74E-05

P-32 1.01E-01

P-33 1.86E-04

S-35 6.39E-03

K-40 5.16E-05

Sc-46 1.30E-06

Cr-51 4.79E-07

Fe-55 6.65E-07

Fe-59 1.53E-06

Mn-54 2.88E-06

Co-56 2.01E-05

Co-57 6.68E-03

Co-58 2.14E-05

Co-60 4.51E-05

Zn-65 9.48E-07

Ni-59 1.33E-06

Ni-63 4.09E-06

Y-88 2.12E-05

Y-91 5.66E-06

Sr-90 2.05E-01
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 35 of 40

ATTACHMENT A

SOURCE TERM BY BUILDING

(Page 9 of 14)

Table A-1 (Continued). Source term by building.a

(Current) building no. Building activity Radionuclide

Activity

fraction

514 Tank Farm Zr-95 1.40E-03

Nb-95 3.08E-05

Mo-99 7.99E-06

Tc-99 2.76E-05

Ru-103 2.54E-06

Ru-106 4.35E-06

Cd-109 8.97E-08

Sn-113 4.66E-09

Ag-110m 8.65E-07

I-125 5.59E-05

I-131 1.06E-05

Sb-120m 7.32E-07

Sb-124 6.25E-06

Sb-125 9.93E-06

Te-132 2.00E-07

Ba-133 4.99E-05

Ba-140 9.99E-07

Cs-134 1.42E-05

Cs-136 9.31E-07

Cs-137 1.08E-01

La-140 2.00E-08

Ce-139 3.99E-09

Ce-141 2.93E-05

Ce-144 1.87E-04

Nd-147 4.46E-06

Pm-147 2.40E-06

Gd-148 1.28E-05

Sm-151 7.99E-08

Eu-152 1.82E-04

Eu-154 1.19E-04

Eu-155 1.48E-05

Eu-156 1.13E-07

Tb-160 1.20E-08

Hf-172 1.93E-06

Hf-181 2.66E-06

Lu-173 1.60E-06

Lu-174 3.33E-07

W-185 4.79E-09

Au-195 4.26E-06

Hg-203 3.06E-20

Bi-207 1.04E-06

Bi-210 5.06E-07

Po-209 4.50E-06

Po-210 5.06E-07

Pb-210 1.36E-05

Ra-226 1.55E-04

Pa-233 1.60E-09

Th-228 6.02E-03

Th-229 1.28E-05

Th-230 1.27E-05

Th-232 1.14E-04

U-232 1.31E-05

U-233 4.44E-05

U-234 1.63E-03

U-235 7.36E-04

U-236 3.33E-07

U-237 3.46E-04
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 36 of 40

ATTACHMENT A

SOURCE TERM BY BUILDING

(Page 10 of 14)

Table A-1 (Continued). Source term by building.a

(Current) building no. Building activity Radionuclide

Activity

fraction

514 Tank Farm U-238 1.02E-02

Pu-236 3.46E-08

Pu-238 1.01E-04

Pu-239 1.50E-03

Pu-240 1.48E-04

Pu-241 3.86E-05

Pu-242 4.37E-05

Pu-244 7.16E-08

Np-237 2.31E-05

Np-239 1.01E-04

Am-241 1.02E-03

Am-243 1.97E-05

Cm-244 4.12E-05

Cf-249 4.63E-06

612 Waste storage and repackaging H-3 9.79E-01

Be-7 1.22E-06

C-14 7.04E-04

Na-22 1.87E-04

P-32 1.29E-02

P-33 1.05E-08

S-35 6.98E-06

Cl-36 3.66E-11

K-40 2.03E-05

Sc-46 1.25E-06

Cr-51 1.34E-06

612 Waste storage and repackaging Fe-55 1.23E-06

Mn-54 4.20E-06

Co-56 7.33E-08

Co-57 5.01E-06

Co-58 1.34E-06

Co-60 1.07E-03

Ni-63 4.89E-08

Zn-65 1.22E-08

Se-75 2.88E-11

Y-88 1.45E-07

Y-91 1.05E-11

Sr-85 8.72E-13

Sr-89 3.93E-12

Sr-90 6.02E-04

Nb-94 5.50E-05

Nb-95 1.34E-06

Zr-90 1.48E-10

Zr-95 1.28E-07

Mo-99 1.55E-09

Tc-99 1.60E-05

Rh-102 2.01E-10

Rh-103 1.66E-08

Rh-103m 4.10E-13

Rh-106 3.49E-11

Ru-106 2.58E-08

Cd-109 4.02E-05

Cd-115 5.25E-09

Ag-110m 2.18E-14

I-125 2.45E-05

I-131 3.15E-09

Sb-124 4.45E-14

Sb-125 4.99E-06
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 37 of 40

ATTACHMENT A

SOURCE TERM BY BUILDING

(Page 11 of 14)

Table A-1 (Continued). Source term by building.a

(Current) building no. Building activity Radionuclide

Activity

fraction

612 Waste storage and repackaging Sm-151 5.10E-11

Ba-133 1.76E-05

Ba-140 7.64E-09

Cs-134 1.58E-05

Cs-137 5.11E-04

Ce-139 2.44E-13

Ce-141 2.85E-08

Ce-144 1.52E-04

Nd-147 1.59E-09

Pm-147 5.95E-10

Pm-151 4.73E-10

Sm-151 8.72E-13

Gd-146 1.05E-05

Gd-148 1.57E-05

Eu-149 2.79E-12

Eu-152 6.28E-05

Eu-154 6.10E-05

Eu-155 5.54E-06

Tb-160 1.74E-13

Hf-172 2.44E-06

Lu-173 1.22E-06

Lu-174 1.22E-06

Ta-182 1.40E-11

W-185 3.66E-09

Ir-192 8.72E-14

Au-195 5.41E-12

Pt-195m 9.59E-10

Hg-203 8.72E-17

Bi-207 1.17E-06

Bi-210 2.44E-07

Po-209 3.30E-06

Po-210 2.44E-07

Pb-210 1.68E-05

Ra-223 4.73E-12

Ra-226 1.72E-06

Ra-228 2.94E-09

Th-228 1.45E-06

Th-229 1.57E-05

Th-230 1.56E-05

Th-232 1.48E-05

Th-234 1.13E-12

U-232 1.58E-05

U-233 2.08E-05

U-234 4.93E-04

U-235 1.85E-04

U-237 3.39E-08

U-238 2.01E-03

U-239 9.60E-12

Pu-238 1.15E-04

Pu-239 3.89E-04

Pu-240 3.54E-05

Pu-241 1.28E-04

Pu-242 2.60E-05

Pu-244 3.56E-06

Np-237 1.69E-05

Np-239 6.09E-07

Am-241 3.74E-04
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 38 of 40

ATTACHMENT A

SOURCE TERM BY BUILDING

(Page 12 of 14)

Table A-1 (Continued). Source term by building.a

(Current) building no. Building activity Radionuclide

Activity

fraction

612 Waste storage and repackaging Am-242 6.81E-09

Am-242m 1.57E-09

Am-243 4.11E-04

Am-244 4.45E-06

Cm-244 1.05E-04

Cf-249 3.39E-06

Cf-250 2.24E-12

612 Yard H-3 1.00E+00

C-14 2.11E-06

P-32 1.61E-07

S-35 3.40E-08

K-40 4.48E-11

Cr-51 4.84E-11

Mn-54 6.11E-13

Co-57 6.38E-13

Co-60 1.19E-10

Ni-59 1.01E-12

Ni-63 2.92E-11

Se-75 2.45E-08

Sr-90 9.84E-12

Nb-95 3.19E-13

Tc-99 2.44E-12

Sb-125 4.09E-12

Cs-134 1.02E-11

Cs-137 3.60E-10

Ce-144 4.25E-10

Pm-147 1.70E-12

Sm-151 5.26E-13

Eu-152 5.45E-10

Eu-154 2.89E-10

Eu-155 6.11E-12

Bi-207 1.81E-13

Bi-214 4.52E-14

Ra-226 3.65E-11

Ra-228 1.22E-11

Th-228 5.50E-10

Th-230 1.17E-14

Th-232 5.69E-10

Th-234 1.17E-11

U-233 5.40E-10

U-234 1.87E-08

U-235 2.47E-09

U-238 1.87E-07

Pu-238 2.43E-09

Pu-239 5.46E-10

Pu-240 2.12E-11

Pu-241 4.43E-09

Pu-242 1.02E-08

Np-239 4.80E-11

Am-241 1.45E-09

Am-242 1.38E-12

Am-243 4.22E-10

Cm-243 4.89E-11

Cm-244 3.35E-13

Cm-245 3.03E-11
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 39 of 40

ATTACHMENT A

SOURCE TERM BY BUILDING

(Page 13 of 14)

Attachment A (Continued). Source term by building.a

(Current) building no. Building activity Radionuclide

Activity

fraction

625 Waste operations H-3 2.14E-01

C-14 2.34E-04

P-32 3.61E-10

K-40 8.78E-05

Mn-54 1.36E-06

Co-57 5.74E-08

Co-60 9.89E-06

Sr-90 2.34E-06

Y-88 1.70E-10

Zr-95 2.55E-06

Ru-106 7.01E-07

Cd-109 4.40E-08

Sb-125 9.78E-07

Ba-133 1.01E-07

Cs-134 1.11E-06

Cs-137 1.05E-05

Ce-141 4.46E-09

Ce-144 1.57E-06

Eu-152 4.47E-06

Eu-154 2.55E-06

Eu-155 2.55E-07

Bi-214 2.98E-08

Pb-212 3.83E-08

Pb-214 4.89E-09

Pa-231 4.89E-08

Ra-226 1.51E-05

Ra-228 6.59E-04

Th-228 1.55E-06

Th-230 5.53E-10

Th-232 3.62E-06

Th-234 4.89E-04

U-233 4.46E-08

U-234 2.76E-01

U-235 4.89E-01

U-238 2.32E-03

U-239 2.55E-07

Pu-238 7.68E-05

Pu-239 3.40E-03

Pu-240 3.83E-06

Pu-241 1.30E-02

Pu-242 4.46E-08

Np-237 2.34E-07

Am-241 2.87E-04

Am-243 5.10E-06

Cm-243 9.78E-09

Cm-244 2.34E-09

Cm-245 1.38E-09

2561 U-234 5.00E-01

U-238 5.00E-01

U-234 8.01E-02

U-235 1.12E-02

U-238 8.66E-01

Site 300-801 H-3 1

Site 300 Pit 7 H-3 1

Site 300-810A U-234 8.38E-02

U-235 1.16E-02

U-238 9.05E-01
Document No. ORAUT-TKBS-0035-5 Revision No. 00 Effective Date: 10/07/2005 Page 40 of 40

ATTACHMENT A

SOURCE TERM BY BUILDING

(Page 14 of 14)

Attachment A (Continued). Source term by building.a

Site 300-810B U-234 8.47E-02

U-235 1.13E-02

U-238 9.04E-01

Site 300-850 H-3 9.6E-01

U-234 3.2E-03

U-235 4.4E-04

U-238 3.4E-02

Site 300-851 H-3 9.57E-01

U-234 9.57E-04

U-235 1.34E-04

U-238 1.02E-02

Site 300 Well 8 Spring H-3 1 

a. This table was derived from data reported for calendar years 1992 through 2001 as part of the National Emission 

Standards for Hazardous Air Pollutants (NESHAPs; 40 CFR Part 61, Subpart H) reporting process (LLNL 1994, 1995, 

1996, 1997, 1998, 1999, 2000, 2001, 2002, 2003). As part of its required reporting under NESHAPs, LLNL must 

monitor emissions by radionuclide and total activity at all building release points. This table only lists buildings for 

which NESHAPS data was reported; this is not a complete list of all buildings. For a listing of major buildings and 

activities see Lawrence Livermore National Laboratory- Site Description (NIOSH, 2005.) While the emissions identified 

over the years might not be fully representative of the source term in each building, they do provide a reasonable basis, 

in the absence of claimant-specific monitoring results, for determining the types of radioactive materials to which 

employees might have been exposed. To develop the table, the isotopic listing of each building in the relevant 

NESHAPs reports was captured in a building-specific spreadsheet. The annual emissions were totaled for each 

isotope and then for all isotopes in the listing. Activity fractions were obtained by dividing the Activity total by the “all 

isotopes” total (Berger 2005). The Activity fraction for each radionuclide in the table is that fraction of the total source 

term present (i.e., the sum of the fractions for each building is equal to “1" or 100% of the mixture). However, those 

radionuclides with Activity fractions equal to “zero”, those with half-lives too short to contribute significant dose (i.e., 

isotopes of nitrogen and oxygen), and those that contribute submersion dose only were subsequently deleted from the 

listing.

b. LLNL (1987a).